Typical fissionable nuclear fuels in metal form comprise uranium, plutonium and thorium, which can be employed alone or alloyed. The fuels currently utilized are primarily comprised of uranium metal. A preferred fuel comprises combinations of a major portion of uranium alloyed with a minor portion of plutonium, for example, 60-98 wt.% uranium metal with 2-40 wt.% plutonium metal.
Fissionable fuel materials in metallic form have a long history in the nuclear reactor field, but due to significant shortcomings, have conventionally been used primarily in the so-called breeder type of reactors utilizing liquid metal coolants. The primary drawbacks of the use of metallic fissionable material as fuel are their relatively low melting temperatures and accompanying loss of structural characteristics such as tensile strength at moderately low temperatures, and their high degree of reactivity with other elements, including susceptibility to corrosion. Uranium in metallic form (the most prevalent fissionable fuel material) melts at only 2070.degree. F. (1132.degree. C.), and an alloyed metal fuel containing uranium and plutonium normally has an even lower melting temperature. For example, an 88 wt.% U--12 wt.% Pu alloy has a melting temperature of about 1130.degree. F. (610.degree. C.).
On the other hand, fissionable fuels in metallic form provide excellent thermal conductivity for highly efficient heat transfer, and maximum concentration of fissionable atoms per unit volume. Thus, more power can be produced per unit volume with metallic fuel and heat can be transferred more efficiently to the liquid coolant.
Alloys of uranium metal and/or plutonium metal have been proposed and used to enhance metal fuels and overcome the shortcomings of metallic fuel. For example, small amounts of alloying metals such as molybdenum, niobium, titanium, zirconium, vanadium or chromium have been used to stabilize phase structures and, in turn, the properties attributable thereto in metallic fuels. For instance, the article "Properties of Uranium-Plutonium-Base Metallic Alloys" by R. J. Dunworth et al., Argonne National Laboratory, Annual Progress Report for 1965, ANL-7155 (1965), pp. 14-25, discloses the alloying of metal fuel with zirconium or titanium to increase the melting temperature of uranium-plutonium fuel.
Zirconium is also included as an alloying component in such metallic fuels to provide an elevated solidus temperature for the fuel and to enhance its chemical compatibility with stainless steel, which is commonly employed in fuel containers for service in liquid metal-cooled nuclear reactors. A preferred metal fuel alloy which has been considered for commercial service comprised an alloyed composition of 64 wt.% U--26 wt.% Pu--10 wt.% Zr. Such an alloy has a melting point of about 1868.degree. F. (1020.degree. C.).
However, additional problems attend the technique of alloying conventional fissionable metal fuels with non-fuel elements such as zirconium metal. It has been found that evidently due to inherent reactor conditions of intense radiation and high temperatures, metal alloy fuels which initially comprise a substantially uniform mixture of alloyed ingredients become chemically redistributed into nonuniform mixtures. This restructuring of the metal alloyed fuel has a pronounced effect upon its properties and their uniformity throughout the mass of the fuel body.
One significant aspect of this redistribution of the alloyed ingredients, such as zirconium, is the inward migration of the zirconium metal to the inner or central area of the fuel body. This is accompanied by an increase in the solidus temperature of the inner or central area and correspondingly a reduced solidus temperature in the outer or peripheral area of the unit. Thus, the melting temperature of the remaining alloy in the peripheral area of the fuel body is lowered, and the intended effect of the added zirconium to avoid low-melting phase formation is reduced or negated. A lower melting temperature of the surface portion of a fuel body increases the potential for chemical interaction with adjoining materials.
When the zirconium migrates, the remaining alloy ingredients form lower-melting-point alloys or eutectic compositions. Moreover, in the absence of zirconium, components remaining in the peripheral area of the fuel, such as plutonium and fission-produced cerium, form low-melting-point phases which can attack or react with the stainless steel of the fuel container (hereinafter "cladding"), thereby degrading the integrity of the cladding. An interaction between fuel components and the stainless steel cladding will degrade the structural strength of the relatively thin wall of the cladding due to reduced thickness, altered composition or resulting permeability.
A paper entitled "Chemical Interaction of Metallic Fuel with Austenitic and Ferritic Stainless Cladding", by G. L. Hofman et al., Argonne National Laboratory, Tucson Conference, September (1986), discussed interdiffusion phenomena between metallic fuel comprising U--Pu--Zr and components of ferritic stainless steel fuel cladding which could degrade the performance of the fuel composition. These phenomena include formation of strength-reducing diffusion zones within stainless steel, intergranular penetration of fuel ingredients into stainless steel, and formation of eutectic areas having a melting temperature below the operating temperature.
Typical liquid metal-cooled nuclear reactors conventionally use stainless steel cladding, preferably those stainless steel alloys marketed under the commercial designations HT9 or D9. Typical compositions for these stainless steel alloys are given in Table 1.
TABLE 1 ______________________________________ Ingredient (wt. %) HT9 D9 ______________________________________ Fe 84.8 65.7 Cr 12.0 14.5 Ni 0.6 14.5 Mo 1.0 2.0 Mn 0.5 2.0 W 0.4 -- V 0.3 -- Si 0.2 1.0 Ti -- 0.3 C 0.2 0.03 N 0.004 0.005 O 0.006 0.006 ______________________________________
Low-melting-point metal alloys form at the interface between U--Zr or U--Pu--Zr metal alloy fuels and stainless steel cladding such as HT9 and D9 alloys. These low-melting-point alloys can seriously degrade cladding performance during high-temperature reactor transient events.
Cladding materials are selected for their resistance to interaction with U--Zr or U--Pu--Zr metal alloy fuels. Nevertheless, diffusion couple studies and irradiation evaluations have demonstrated that diffusion of plutonium, uranium and fission products to the cladding material occurs without a corresponding diffusion of enough zirconium to maintain the desired high-melting-point alloy composition at the interface of the fuel and the cladding.
Irradiation studies demonstrated that contact between HT9 cladding and 71 wt.% U--19 wt.% Pu--10 wt.% Zr metal alloy fuel resulted in a 7 to 10--.mu.m-deep reaction zone in the steel alloy at 2.9 a/o burnup, which zone was enriched in plutonium and the radiation product cesium. Both plutonium and cesium form low-melting-point alloys with stainless steel alloy constituents as illustrated in Table 2.
TABLE 2 ______________________________________ Melting Point Composition Alloy (.degree.C.) (Wt. %) ______________________________________ Pu-Fe 410 2.5 Fe Pu-Ni 475 4 Ni Ce-Ni 477 8 Ni Ce-Fe 592 4 Fe Pu-U 610 12 U Pu-Mo 613 0.5 Mo Pu-Cr 615 0.5 Cr Pu-Ce 625 12 Ce Pu 640 -- Pu-Zr &gt;640 * U-Fe 725 11 Fe U-Ni 740 11 Ni Zr-Ce 750 3 Zr U-Cr 860 5 Cr U-26Pu-2Zr 913 -- Zr-Fe 934 16 Fe Zr-Fe 961 17 Ni U 1134 -- U-Zr &gt;1134 * Zr-Cr 1300 18 Cr ______________________________________ *Form solid solutions, Pu and U have lowest melting points in the system.
The nonuniform melting conditions of metal alloy fuels in reactor service and the potential effect is discussed in an article entitled "Post-Irradiation Examination of U--Pu--Zr Fuel Elements Irradiated in EBR-11 to 4.5 Atomic Percent Burnup" by W. F. Murphy et al., Argonne National Laboratory, ANL-7602, November (1969). This article additionally discusses the extensive physical changes which occur in metal alloy fuel during fission, such as extensive deformation. Such deformation includes expansion or swelling of up to about 30% by volume due to thermal effects and internal generation of fission-produced gases.
The potential for cladding failure is discussed in an article entitled "Metallic Fuel Cladding Eutectic Formation During Post-Irradiation Heating" by B. R. Seidel, Argonne National Laboratory, Trans. Ans. 34, June (1980), pp. 210 and 211.
A method of inhibiting an interaction between a metal alloy fissionable fuel for a nuclear reactor and a stainless steel cladding for the fuel was disclosed in U.S. Pat. No. 4,971,753, which patent is assigned to the assignee of the present application. In accordance with that teaching, a fuel element having a body of alloyed uranium metal fuel housed within stainless steel cladding is provided with an expendable source of alloying metal which raises the melting temperature of the metallic fuel. In particular, a zirconium barrier is placed between the alloyed uranium metal fuel body and the stainless steel cladding. This barrier protects the stainless steel cladding from the deleterious effects of low-melting-point alloys resulting from irradiation by increasing the availability of zirconium to raise the melting point of alloys formed at the fuel/cladding interface.
U.S. Pat. No. 4,971,753 states that the zirconium barrier can be installed in the fuel element in the form of a layer of zirconium formed on the outer peripheral surface of the fuel body or on the inner peripheral surface of the cladding; or in the form of a hollow circular cylinder, e.g., rolled zirconium metal foil or a rigid tubular section, of zirconium inserted in the space intermediate the fuel body and cladding.
The use of barrier material of circular cross section between the metal fuel slugs and the stainless steel cladding of the fuel pin divides the gap between the fuel and the cladding into two smaller gaps: one between barrier and fuel and the other between barrier and cladding. Thus, one result of barrier insertion is that a second gap is created between the fuel inside the cladding and reactor coolant outside the cladding. Another result is that the diametrical gaps between the various pin components are reduced. These gaps must be filled with liquid sodium (or other suitable liquid metal), that is, the components must be sodium thermal bonded, to provide an acceptable heat transfer path between the fuel and reactor coolant and thereby prevent unacceptable temperatures during operation.
The top of the space between the fuel and cladding is typically occupied by helium gas. If the respective opposing surfaces of pin components form a gap so small that those surfaces are not wetted by the liquid sodium and the gap is filled with gas bubbles, then the sodium thermal bond will have a void. Excessive voiding in the sodium thermal bond could produce a temperature differential between the fuel and cladding of 500.degree.-600.degree. C., causing localized fuel melting during radiation. Thus, it is critical that the quality of the sodium thermal bond be verified during fabrication of the fuel pins.
In an article by A. L. Pitner and J. O. Dittmer entitled "FFTF Metal Fuel Pin Sodium Bond Quality", Trans. Am. Nucl. Soc., Vol. 59, pp. 144-145 (1989), the results of sodium bond quality tests for metal fuel pins without barriers were reported. The contents of this article are incorporated by reference herein.
In this article, the authors concluded that small voids in the sodium bond encountered in the lower regions of the fuel pin were invariably situated on the side where an off-center fuel slug contacted the cladding. The high surface tension of the liquid sodium prevented wetting of the small gap areas. Thermal analysis of this metal fuel pin design revealed that a significant fuel temperature increase was not observed until the unbonded area, i.e., the so-called "degree of circumferential unbonding", approached an angle of 90 degrees around the cladding circumference.
This teaching has applicability to fuel pins with barriers. The of-center location of the fuel slug results from a combination of gravity and divergence from slug linearity. The use of a circular, thin-walled barrier tube between the metal fuel and cladding will result in a fuel pin cross section in which the off-center fuel slug and barrier are located on one side of the pin, with the barrier sandwiched between the fuel slug and cladding. The addition of the barrier tube results in smaller gaps between the fuel pin components which provide less diametrical area for sodium bonding. As a result, the potential for unbonded areas in the barrier fuel pin is increased.